The Shanghai Nuclear Engineering Research and Design Institute (SNERDI) designed the CNP-300 for the CNNC. The design is a down-sized traditional two — loop pressurized water reactor design. The core consists of 121 fuel assemblies with a 15 X 15 array of zircaloy clad fuel pins. Two vertically mounted external reactor coolant pumps circulate the primary coolant between the reactor pressure vessel and the two vertical U-tube steam generators. Although initially designed for 300 MWe, subsequent units were upgraded in capacity to 325 and 340 MWe. Because the CNP — 300 uses a loop-type configuration, a large-break loss of coolant accident is possible and multiple safety systems are incorporated to mitigate its consequences, including high-pressure injection systems.

The first CNP-300 was constructed by the CNNC at the Qinshan Nuclear Power

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Plant site in China and was connected to the grid in 1991. Two subsequent units, Chashma 1 and 2, were constructed in Pakistan. Chashma 1 commenced operation in 2000. Units 3 and 4 are currently in design and construction phases. Key parameters and a representative graphic for the CNP-300 design are given in Figure 2.3. [3]

Small reactors and the modular construction of reactors are not new. Historically, early reactors for commercial production of electricity were of small size, a consequence of the prudent engineering process of constructing plants starting at small ratings to gain the needed construction and operating experience necessary to move confidently to larger ratings. Now, after a half-century of experience, commercial civil reactors are being deployed with ratings up to 1660 MWe. Additionally, small units were built for terrestrial deployment to provide electric power for remote, vulnerable military sites; for ocean deployment for propulsion of submarines, naval and commercial ships and for aircraft propulsion. Modular construction techniques historically have also been used for serial production of selected products. However, what is new is the vision of small rated power reactors composed of a single or multiple modules sized to markets of small — or large-sized electric grids, thereby creating new nuclear generating sites which require significantly reduced capital investments and capital investment rates. The further economic premise is that electric generation cost can be made sufficiently comparable to that of existing large-sized plants by employing a strategy of economy of numbers (manufacture of multiple identical modules) and simplification of design versus the traditional economy of scale.

Canada was the first country to develop commercial power reactors based on heavy — water coolant. The heavy water, i. e. water in which the two hydrogen atoms are replaced by deuterium atoms, is an attractive coolant and moderator because it has a much lower tendency to absorb neutrons. This allows the reactor to be fueled by natural uranium rather than requiring the complex and expensive process of enriching the uranium. Because of the excellent ‘neutron economy’ provided by the low neutron absorption of the heavy water, this coolant became a favorite option for several ‘production’ reactors that were built by the United States and the former Soviet Union for the purpose of producing weapon-usable fissile material.

India imported the Canadian ‘ CANDU’ technology and developed an indigenous design of a 220 MWe plant. Although India subsequently up-scaled their plant design to capacities of 540 and 700 MWe, they continue to build 220 MWe plants. Because of the limited uranium resources but abundant thorium resources in India, they have developed a three-stage strategy for achieving a self-sustaining power production capability. The first stage utilizes the pressurized heavy-water reactors (PHWRs) (220, 540 and 700 versions) to generate sufficient plutonium

to support a fleet of fast spectrum reactors, which constitutes the second stage. The sodium-cooled fast reactors will include blankets of fertile thorium to produce 233U, which provides the feedstock for the third stage: advanced heavy-water reactors AHWRs. At equilibrium, the AHWRs will be self-sustaining and produce sufficient 233U from the U-Th or Pu-Th mixed oxide fuel to support subsequent fuel loads.

Table 2.2 provides a summary of the two available or planned SMR designs based on heavy water technology.

The economic characteristics of large water power reactors are known from years of construction and operating experience. The cost of sodium-cooled reactors based on deployments of demonstration units in the late 1900s has led to capital cost estimates of 110-125% that of water-cooled reactors (Waltar et al., 2012). Experience with gas-cooled, and certainly lead/lead-bismuth-cooled, reactors has not been sufficient to allow a comparable projection of overnight capital costs compared to water-cooled reactor experience. Hence, while it is accepted that the capital cost of individual SMR units will be far lower than that of the large-rated reactors employing the same coolant, the capital cost per KWe for SMRs compared to large-rated reactors, although likely larger, is as yet not established. We can only project comparative costs of SMRs employing the various coolants on the basis of the above-noted large-rated reactor experience.

Other potential measures of comparative economic characteristics of variously cooled SMRs are the fundamental parameters of core power density and specific power. The power density, kW/liter, reflects the core volume and hence is often a measure of the vessel containment and plant size necessary for a given power rating. Exceptions do exist if the reactor vessel or containment size is dictated by considerations other than core power density. For example, the SPRISM sodium — cooled fast reactor vessel is sized to accommodate decay heat removal through an air-cooled chimney outside the guard vessel: BWR containments by virtue of their use of in-containment coolant pools for pressure suppression are much smaller than those of PWRs, which control pressure by large air-filled containment volume. The power density is thus a relative indication of capital cost, albeit for plants using comparable design strategies and principal materials. The specific power, kW/kgIHM, reflects the mass of initial heavy metal (IHM) or fuel needed for a given power rating. The specific power is thus a relative indication of fuel cycle cost, but for plants using comparable fuels.

However, it is clear that not all SMRs employing the various coolants of interest use comparable materials or fuels. Hence the relative values of power density and specific power presented in Table 1.5 for various coolants do not necessarily forecast the comparative economic character of reactors employing various coolants. Nevertheless, these parameters provide an insight regarding the significant benefit to sodium-cooled reactors from their high relative parametric values, a benefit which

likely keeps their costs close to water-cooled designs even though they use an exotic liquid metal coolant requiring considerable costly instrumentation and purification systems and their enrichment is much higher than that of water-cooled designs. Furthermore, the low parametric values of the helium-cooled reactor indicate the inherent economic disadvantage of large reactor volume which this reactor coolant type faces. However, unless one considers all aspects of the design, surrogate parameters of cost can be misleading. What is needed is an integrated cost analysis to include the design of the reactor system, all needed safety systems, the power conversion system considering thermal efficiency, operating staff size, maintenance cost, and fuel costs to evaluate the economic competitiveness of any design as measured in cents/kW h of power produced.

Nuclear power plants must:

• be economically competitive;

• have superior safety; and

• satisfactorily manage their waste.

These are the three imperatives, which are shared by all power plants, but have a particular importance, especially the last two, in the case of nuclear plants.

Economic competiveness is an obvious given. It is always a necessary condition, but in the case of nuclear power it is not sufficient, as it must be accompanied by the other two. Also, it must be substantially better than the competition and bring in additional considerations to compensate for the mere fact of being nuclear.

Safety must be far superior to conventional power plants because the consequences of a nuclear accident might impact a much larger population because of the dispersion of radioactive effluents. How critical safety is for acceptance of nuclear plants has been underscored by the aftermath of the three major accidents which have occurred. Objectively speaking, Three Mile Island (1979) was actually a minor accident, with very moderate radiation release, but it became the catalyst to halt nuclear growth for more than a decade. In the meantime came Chernobyl (1986) which was eventually overcome because of the atypical conditions of this Soviet reactor. When nuclear was on the rebound, Fukushima (2011) happened and several countries recoiled from nuclear, the most striking example being Germany, which is very willingly accepting the economic penalty of forgoing its nuclear plants.

Disposal of the waste is on a completely different scale for nuclear power than for any other power source. For the latter, handling of the waste is a minor, or at least manageable, issue. For nuclear plants, even though various technical solutions are available to deal with their waste, the mantra is that nuclear waste ‘will poison the earth for millennia’.

So, where does nuclear power stand now in fulfilling the three imperatives?

• Economics: green light, mostly. It is a highly competitive field, but the bottom line is that hundreds of nuclear power plants are operating and no utility would be willing to spend billions of dollars to build a plant which is not economically competitive.

• Safety: yellow light. Nuclear plant safety has proven to be fundamentally sound; older plants have been improved or shut down and new designs have much improved safety. The key issue is the eventual occurrence of ‘big’ accidents.

• Waste disposal: flashing red light. Resolution of the waste problem has not made any substantial progress from the early days of nuclear power and disposal at site. Technical solutions do exist, but the problem is political.

Small modular reactors (SMRs) can satisfactorily address the three imperatives. Moving in reverse, in regards to the waste disposal imperative, fast spectrum reactors run in a burner mode will dispose of the plutonium and minor actinides. Large and smaller fast and thermal reactors using both uranium and thorium cycles can significantly reduce the present waste legacy as well as avoid future additions.

To turn this third imperative from red to green requires the political will to move ahead with available technological solutions.

The scenario regarding the safety imperative is completely different. The iPWR is the very type of plant uniquely capable of reaching the ultimate safety in a reasonably short time using extensively proven technology.

The basic tenet of nuclear safety is very simple. The plant must be capable, through its intrinsic design and auxiliary safety systems, to survive any conceivable accident without releasing excessive radiation. Of course the rub is in the definition of ‘conceivable accident’ and ‘excessive radiation’, as well as the adopted design and the choice of auxiliary safety systems.

Major accident probabilities are defined through the core damage frequency (CDF), that is the probability that a postulated accident results in core damage, which is automatically considered as cause for radiation release to the environment. The CDF target for the early reactor designs was of the order of E-4/yr (0.0001 events/ yr), that is the probability for core damage and radiation release, accounting for all hypothetical accidents, was once in 10 000 years; very small indeed, especially considering that the design lifetime was 30 years. However, the number of plants kept increasing, as well as their operating life. If the probability of core damage remained at E-4/yr for every plant and there are, say, 400 plants around the world with an average lifetime of 30 years, the probability of a major accident at any plant in the world would be 400 E-4 or 0.04 every single year, 1.2 over a 30-year lifetime, which is approximately the Three Mile Island/Chernobyl time frame. After approximately another 30 years, Fukushima. The pattern is there.

It is almost certain that another major accident in the next 15-30 years will be the end of the line for nuclear power. It is therefore necessary to decrease the CDF value. After Three Mile Island, a very significant amount of effort was spent to increase the reliability and redundancy of the safety systems. This improved the safety, but also significantly increased the reactors’ price tag. No wonder that no new nuclear plants were ordered for quite some time. A breakthrough occurred with the adoption of passive safety systems in lieu of the previously active systems. The CDF for the new designs dropped to the order of E-6; also, existing designs are being retrofitted to decrease their CDF. However, the lifetime of the new designs as well as of the retrofitted ones is now projected to be of the order of 60 years. Finally, the new passive designs with a CDF of E-6 have a very hefty price tag and thus the market tendency is to prop up the older designs for as long as possible.

A clean breakthrough is necessary, i. e. to introduce reactors having a very low CDF as well as a low price tag to easily replace the old plants.

First, the CDF. If the CDF probability is decreased to E-8, which is the frequency of a severe event known as ‘act of God’, 1000 plants with a lifetime of 60 years will yield a failure probability over their 60 year lifetime of 6E-4, or 6 major accidents over 10 000 years. Even with 10 000 nuclear plants worldwide, we would have major accidents at an average interval of 170 years. The safety issue has disappeared.

Is it possible to have nuclear plants with such an infinitesimal CDF value? For the current generations of plants operating, in construction or offered now, the answer is no. With a new generation of plants, specifically SMRs of the iPWR type presented in Section 3.3, the answer is possibly yes. It becomes a definite yes if the iPWR design is safety-driven, as will be discussed in Section 3.4. Finally, this safety driven iPWR must be economically competitive to cover a majority of the market; this will be discussed in Section 3.5.

The Flexblue SMR concept is one of the most recent and more unique entries into the commercial SMR competition. Based substantially on submarine experience, the SMR is intended to be operated on the sea floor at a depth between 50 and 100 m. The 50 m upper limit isolates the plant from surface storm effects while the 100 m lower limit protects against excessive pressure on the 14 m diameter by 146 m long outer hull. The 160 MWe SMR can be deployed in multi-unit ‘farms’ that are remotely operated from a coastal command and control facility. A heavy lift cargo ship transports the module to and from its operating site. Refueling is expected to occur every two or three years, at which time it will be returned to a central facility for refueling and maintenance. Later versions of the transport ship may allow for refueling at the site of deployment. Major overhauls of the module are expected every 10 years. At the end of the module’s life, expected to be 60 years, the module is transported to a decommissioning facility for final disposal.

The ocean environment greatly simplifies traditional safety aspects of the design. The ocean water provides an assured ultimate heat sink and an additional barrier to fission products in the case of an accident. The passive safety systems ensure that

the module can survive an unlimited time without operator action or power following an accident. Key parameters and a representative graphic for the Flexblue design are given in Figure 2.4. [4]

Commercial electric power began with small reactors of light-water-cooled design. Key examples are the Shippingport, 60 MWe reactor designed by the Westinghouse — operated Bettis Naval Atomic Power Laboratory, which started operation in 1958; the Yankee Rowe reactor, 185 MWe (Westinghouse) in 1960; the Indian Point One reactor, 275 MWe (B&W) in 1962 (all pressurized water reactor [PWR] designs); and Dresden 210 MWe (General Electric) in 1960 (a boiling-water reactor (BWR) design).

The eight military reactors for terrestrial application developed by the US Army Nuclear Power Program included (1) the stationary plants operated at Fort Belvoir, Virginia, which started operation in April, 1957, seven months before Shippingport and five years before criticality of the Ft. Greely, Alaska reactor; (2) the portable reactor operated at McMurdo Sound at the South Pole in 1962; and (3) a barge — mounted reactor operated off the coast of Panama City, Panama, in 1967. These plants ranged from 1.75 to 10 MWe and performed either a heating or desalinization function in addition to the generation of electricity. Another example of a portable reactor is the Russian PAMIR reactor designed primarily to power remote military radar outposts. The first was the TES-3, a 2 MWe nuclear plant completed in 1961. The design was modified in the 1980s to a smaller, more mobile 630 kW reactor.

The much larger US naval program, which pioneered the application of nuclear power for the propulsion of submarines and surface ships, has produced multiple pressurized water reactors and one sodium-cooled reactor of small ratings. Additionally, several countries have followed suit with naval propulsion — most notably Russia, which expanded its development of water-cooled submarine reactors to submarines using lead-bismuth coolant and has also built nuclear powered naval surface ships and ice-breakers.

Commercial (merchant marine) propulsion has also been exploited through the development of ocean freighters and icebreakers. Four freighters, all with reactors of light-water design, have been built and operated albeit without commercial success: (1) the US Savannah, 74 MWt, in effective service starting 1962; (2) the German Otto Hahn, 38 MWt, 1968; (3) the Japanese Mutsu, 36 MWt, 1972; and (4) the only vessel still in operation under nuclear power, the Russian Sevmorput, 135 MWt, delivered in 1988, which also has ice-breaking capability.

The Otto Hahn reactor design is of special interest since its integral design characteristic is the typical configuration being exploited by several modern PWR SMR vendors. As extensively elaborated in Chapter 3, the term integral design means the co-location of all components and piping of the primary coolant system in the single pressure vessel. By contrast, the typical large-rated PWRs are loop systems with the primary system components, e. g., the steam generators, primary coolant pumps and pressurizer connected by piping to each other, and the pressure vessel which houses the reactor core and the control elements.

To date Russia alone has constructed and operated nine nuclear-powered ice­breakers, starting in 1959 with the Lenin. Two vessel classes have been built: the Arktika class, each vessel with two OK-900A reactors each of 171 MWt; and the Taymy class, each vessel with a single KLT-40M reactor of 135 MWt. (NB: All reactors of the ocean vessels noted above drive propulsion shafts, thus their ratings are only in MWt.) Russia is also constructing a non-self-propelled floating nuclear power station, the Akademik Lomonosov, to provide power supply to remote coastal towns. The reactor station scheduled for delivery in 2016 is powered by two modified ice-breaker reactors, each a KLT-40S reactor of 35 MWe. With these reactors the station can provide either 70 MWe of power, 300 MWt of district heating or 240 000 m3/day of fresh water.

The development of a nuclear propulsion system for military aircraft was initiated in 1946 as the US Nuclear Energy for the Propulsion of Aircraft (NEPA) project and continued under the name of the Aircraft Nuclear Propulsion (ANP) program. Two different systems for nuclear-powered jet engines were pursued — a direct air cycle concept developed by General Electric and an indirect air cycle by Pratt & Whitney. Only the direct air cycle program advanced sufficiently to produce reactors. The first product of the GE program was the Aircraft Reactor Experiment (ARE) which operated for 1000 hours in 1954. It was a 2.5 MWt nuclear reactor experiment using molten fluoride salt (NaF-ZrF4-UF4) as fuel, a beryllium oxide (BeO) moderator, and liquid sodium as a secondary coolant. In 1955, this program produced the successful X-39 engine with heat supplied by the Heat Transfer Reactor Experiment-1 (HTRE-1). The HTRE-1 was replaced by the HTRE-2 and eventually the HTRE-3 unit powering the two jet turbines. Additionally, an operating reactor named the aircraft shield test reactor (ASTR), was flown aboard a modified B-36 bomber to test shielding rather than powering the plane. The HTRE-3 used a shield system of flight-type design but was not taken to power before the program was canceled in 1961.

All these earlier reactors led to the current explicit offering of reduced size modular power plants. These current SMRs, listed in Table 1.1, encompass all coolant technologies being exploited for larger nuclear reactors. Table 1.1 lists only

those reactors offered by international industries. Additional reactors not included in Table 1.1 are under development by national research institutions but have not yet reached the commercialization stage. For example, the fluoride-salt-cooled high — temperature reactor (FHR) (Forsberg et al., 2013) is a 180 MWe reactor with 700 °C peak operating temperature coupled to an air-Brayton combined cycle system.

The Pressurized Heavy Water Reactor (PHWR-220) is a 235 MWe pressure-tube type reactor derived from the Canadian CANDU design. It uses heavy water (D2O) both as a primary coolant and as a neutron moderator. Pressure tubes made of Zr — 2.5%Nb contain the 0.5 m long cluster-type fuel bundles containing 19 Zircaloy-4 cladded pins of natural UO2. Each pressure tube contains 12 fuel bundles. A total of 306 horizontal pressure tubes and surrounding moderator make up the core, which is surrounded by a light-water calandria vessel. Unlike most reactors that use batch refueling, the PHWR is refueled on a continuous basis using two refueling machines — one on either end of the core. Refueling is accomplished by inserting a fresh fuel bundle into one end of the pressure tube and collecting the spent fuel bundle that is forced out on the other end, which is then transported to a spent fuel area. Refueling is bi-directional with eight bundles being refueled at a time. Four reactor coolant pumps are used to circulate the primary coolant through two independent primary heat transport circuits, each providing coolant to 153 pressure tubes. Two recirculating U-tube steam generators on each primary heat transport circuit transfer heat to the secondary side for power conversion.

A double containment system is used for each reactor with an interior pre-stressed concrete structure surrounded by a reinforced concrete structure. The negative pressure gap between the two structures facilitates leak detection.

India currently operates 16 PHWR-220 units at five different locations. The reactors are typically deployed as twin units with some shared systems. Key parameters for the PHWR-220 design are given in Figure 2.14. [12]

Key parameters Electrical capacity: 235 MWe Thermal capacity: 755 MWt

Configuration: Pressure tube

Primary coolant: Heavy water

Moderator: Heavy water

Primary circulation: Forced Outlet temperature: 293 °C RV diameter/height: 6.0 m/4.2m Steam generator: Recirculating

U-tube (4)

Power conversion: Indirect Rankine

Fuel (enrichment): UO2 (< 5%)

Reactivity control: Rods

Refueling cycle: Continuous

Design life: 40 years

Status: 16 units in operation

Figure 2.14 PHWR-220 (India) — Nuclear Power Corporation of India Ltd (NPCIL).

2.3.1 India: AHWR-300-LEU design

The Advanced Heavy Water Reactor (AHWR-300-LEU) being developed by the BhaBha Atomic Research Center (BARC) is an extension of the PHWR-220 with

Gravity driven
water pool

Steam drum Refueling machine

Vertical calandria

Figure 2.15 AHWR-300-LEU (India) — BhaBha Atomic Research Center (BARC).

several significant differences. Although it retains use of heavy water for neutron moderation, it uses light water as the primary coolant. Fuel bundles consist of 54 pins of cladded mixed-oxide fuel, including 30 thorium-uranium oxide pins and 24 plutonium-thorium oxide pins. The fuel bundles are contained in vertical pressure tube channels. The U and Pu content in the mixed-oxide fuel is maintained below 5%. Natural circulation of the primary coolant is used with a direct Rankine cycle for power conversion. The steam/water mixture that exits the core is circulated to an external steam separator drum where the steam is directed to the turbine and the water condensate is returned to the feedwater header.

Currently it is intended that the AHWR-300-LEU will be used in co-generation mode and produce 2400 m3/day of potable water by extracting a portion of the steam from the low-pressure turbine. Key parameters and a representative graphic for the AHWR-300-LEU design are given in Figure 2.15. [13]

All SMRs will be designed to meet the same top level set of regulatory requirements. However, the inherent characteristics of each coolant significantly influence the means by which such requirements are achieved.

Neutronic-based coolant density coefficients of reactivity are of sufficient magnitude as to affect the design of all liquid-cooled reactors. For PWRs and BWRs the moderator coefficient is designed to remain negative for accident conditions, but for sodium — and lead/lead-bismuth eutectic-cooled fast reactors the reactivity coefficients which would be present in significant coolant voiding events are unavoidably positive and protected by other design features. Helium-cooled reactors of either the pebble or prismatic type have the unique feature of low power density coupled with a high heat capacity core and reflector that yields a design such that the reactor, upon an increase in temperature, neutronically reduces the power to a very low level, i. e., well below 1% of full power.

The inherent means to cope with severe accident conditions and design basis events are key features of the variously cooled SMRs, namely,

• bounding the potential energy release.

• mitigation of release of fission products by scrubbing in primary coolants.

• response to the loss of coolant accident (LOCA) and provision for ultimate removal of decay heat.

Of the three imperatives (economics, safety, waste disposal) the iPWR does not bring any new approach to the third one, since it is a PWR. Thus, the waste management aspect is not discussed any further here.

Regarding the first imperative, the SMRs in general do not appear at first to be the logical design for achieving competitive economics, since they go against the economy of scale that has prompted larger and larger plants. A new type of design therefore needs to be developed, one that is simpler, requiring fewer components, and that can be built in a shorter time than the present light-water reactors (LWRs). Here the introduction of the iPWR, which capitalizes on the multi-decade development, construction and operating experience of the PWR, is a plant representing the vast majority of nuclear plants built to date. The iPWR designs proposed and investigated do indeed promise economic competiveness with present nuclear plants as well as conventional energy sources.

However, the raison d’etre of the iPWR is how it aggressively and innovatively addresses the second imperative, superior safety. The configuration has intrinsically this capability, for example by eliminating the occurrence of large LOCAs (loss of coolant accidents) through locating the steam generators inside the vessel, or of control rod ejection accidents, through also locating the control rod drive mechanisms inside the vessel. Most remarkably, the iPWR has the capability, if properly designed, to address synergistically the first two imperatives, that is an increase in safety is concomitant with a decrease in cost, as will be elaborated later in Section 3.5.